Technological problems anticipated in the application of fusion reactors to space propulsion and power generation
Technological problems anticipated in fusion reactor application to spacecraft propulsion and power generation
Rayle, W. D.↗
SEARCH · Search NASA
Results for “fusion reactors”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Helium-3 blankets for tritium breeding in fusion reactors
It is concluded that He-3 blankets offers considerable promise for tritium breeding in fusion reactors: good breeding potential, low operational risk, and attractive safety features. The availability of He-3 resources is the key issue for this concept. There is sufficient He-3 from decay of military stockpiles to meet the International Thermonuclear Experimental Reactor needs. Extraterrestrial sources of He-3 would be required for a fusion power economy.
Steiner, Don↗
Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review
Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ~400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α'), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Accuracy and convergence of coupled finite-volume/Monte Carlo codes for plasma edge simulations of nuclear fusion reactors
The plasma and neutral transport in the plasma edge of a nuclear fusion reactor is usually simulated using coupled finite volume (FV)/Monte Carlo (MC) codes. However, under conditions of future reactors like ITER and DEMO, convergence issues become apparent. This paper examines the convergence behaviour and the numerical error contributions with a simplified FV/MC model for three coupling techniques: Correlated Sampling, Random Noise and Robbins Monro. Also, practical procedures to estimate the errors in complex codes are proposed. Moreover, first results with more complex models show that an order of magnitude speedup can be achieved without any loss in accuracy by making use of averaging in the Random Noise coupling technique.
71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
Proton Collimators for Fusion Reactors
Proton collimators have been proposed for incorporation into inertial-electrostatic-confinement (IEC) fusion reactors. Such reactors have been envisioned as thrusters and sources of electric power for spacecraft and as sources of energetic protons in commercial ion-beam applications.
Miley, George H.↗
Optimisation of confinement in a fusion reactor using a nonlinear turbulence model
We report the confinement of heat in the core of a magnetic fusion reactor is optimised using a multidimensional optimisation algorithm. For the first time in such a study, the loss of heat due to turbulence is modelled at every stage using first-principles nonlinear simulations which accurately capture the turbulent cascade and large-scale zonal flows. The simulations utilise a novel approach, with gyrofluid treatment of the small-scale drift waves and gyrokinetic treatment of the large-scale zonal flows. A simple near-circular equilibrium with standard parameters is chosen as the initial condition. The figure of merit, fusion power per unit volume, is calculated, and then two control parameters, the elongation and triangularity of the outer flux surface, are varied, with the algorithm seeking to optimise the chosen figure of merit. A twofold increase in the plasma power per unit volume is achieved by moving to higher elongation and strongly negative triangularity.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DUCTILE-PHASE TOUGHENED TUNGSTEN FOR PLASMA-FACING MATERIALS IN FUSION REACTORS
Tungsten (W) and W-alloys are the leading candidates for plasma-facing components in nuclear fusion reactor designs because of their high melting point, strength retention at high temperatures, high thermal conductivity, and low sputtering yield. However, tungsten is brittle and does not exhibit the required fracture toughness for licensing in nuclear applications. A promising approach to increasing fracture toughness of W-alloys is by ductile-phase toughening (DPT). In this method, a ductile phase is included in a brittle matrix to prevent on inhibit crack propagation by crack blunting, crack bridging, crack deflection, and crack branching. Model examples of DPT tungsten are explored in this study, including W-Cu and W-Ni-Fe powder product composites. Three-point and four-point notched and/or pre-cracked bend samples were tested at several strain rates and temperatures to help understand deformation, cracking, and toughening in these materials. Data from these tests are used for developing and calibrating crack-bridging models. Finite element damage mechanics models are introduced as a modeling method that appears to capture the complexity of crack growth in these materials.
Ductile Phase Toughening↗
Suppression of irradiation hardening in tungsten-coated ferritic steel for fusion reactor blanket applications
W-coated reduced activation ferritic steels have been developed for use as plasma facing components in fusion reactor blankets, offering excellent sputtering resistance and structural strength. Previous high-temperature coating methods, such as diffusion bonding and brazing, caused interfacial deterioration due to thermal stress from mismatched thermal expansion between W and reduced activation ferritic steel. To address this, underwater explosive welding was introduced as a high-velocity cold process that joins dissimilar materials while maintaining a strong, thin interface without the thermal issues associated with traditional methods. In this study, the effects of neutron irradiation on the hardness and microstructure in W-coated F82H reduced activation ferritic steel (W/F82H) joined by underwater explosive welding are investigated. Following neutron irradiation at 290 °C, irradiation hardening is suppressed in W, F82H, and their interface within the W/F82H material. Furthermore, microstructural observations indicate that the recovery of work hardening and relaxation of elastic strain introduced during coating significantly contribute to the suppression of irradiation hardening in W/F82H. In conclusion, W/F82H exhibits significantly suppressed irradiation hardening compared with those in stand-alone materials. This suppression is explained by residual stress from thermal expansion mismatch and the unique microstructure at the interface. These results provide valuable insights for the development of more durable materials in nuclear fusion applications.
36 MATERIALS SCIENCE↗
Atomistic simulations of deuterium irradiation on iron-based alloys in future fusion reactors
Iron-based alloys are now being considered as plasma-facing materials for the first wall of future fusion reactors. Therefore, the iron (Fe) and carbon (C) erosion will play a key role in predicting the life-time and viability of reactors with steel walls. In this work, the surface erosion and morphology changes due to deuterium (D) irradiation in pure Fe, Fe with 1% C impurity and the cementite, are studied using molecular dynamics (MD) simulations, varying surface temperature and impact energy. The sputtering yields for both Fe and C were found to increase with incoming energy. In iron carbide, C sputtering was preferential to Fe and the deuterium was mainly trapped as D 2 in bubbles, while mostly atomic D was present in Fe and Fe–1%C. The sputtering yields obtained from MD were compared to SDTrimSP yields. At lower impact energies, the sputtering mechanism was of both physical and chemical origin, while at higher energies (>100 eV) the physical sputtering dominated.
36 MATERIALS SCIENCE↗
Overview of the KSTAR experiments toward fusion reactor
The Korean Superconducting Tokamak Advanced Research has been focused on exploring the key physics and engineering issues for future fusion reactors by demonstrating the long pulse operation of high beta steady-state discharge. Advanced scenarios are being developed with the goal for steady-state operation, and significant progress has been made in high ℓ i , hybrid and high beta scenarios with β N of 3. In the new operation scenario called fast ion regulated enhanced (FIRE), fast ions play an essential role in confinement enhancement. GK simulations show a significant reduction of the thermal energy flux when the thermal ion fraction decreases and the main ion density gradient is reversed by the fast ions in FIRE mode. Optimization of 3D magnetic field techniques, including adaptive control and real-time machine learning control algorithm, enabled long-pulse operation and high-performance ELM-suppressed discharge. Symmetric multiple shattered pellet injections (SPIs) and real-time disruption event characterization and forecasting are being performed to mitigate and avoid the disruptions associated with high-performance, long-pulse ITER-like scenarios. Finally, the near-term research plan will be addressed with the actively cooled tungsten divertor, a major upgrade of the NBI and helicon current drive heating, and transition to a full metallic wall.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
System and method for small, clean, steady-state fusion reactors
According to some embodiments, a system for widening and densifying a scrape-off layer (SOL) in a field reversed configuration (FRC) fusion reactor is disclosed. The system includes a gas box at one end of the reactor including a gas inlet system and walls of suitable heat bearing materials. The system further includes an exit orifice adjoining the gas box, wherein the exit orifice has a controllable radius and length to allow plasma to flow out from the gas box to populate the SOL with the plasma. The system may also include fusion products, which decrease in speed in the plasma in the SOL, allowing energy to be extracted and converted into thrust or electrical power and further allowing ash to be extracted to reduce neutron emissions and maintain high, steady-state fusion power.
Cohen, Samuel A.↗
System and method for small, clean, steady-state fusion reactors
According to some embodiments, a system for widening and densifying a scrape-off layer (SOL) in a field reversed configuration (FRC) fusion reactor is disclosed. The system includes a gas box at one end of the reactor including a gas inlet system and walls of suitable heat bearing materials. The system further includes an exit orifice adjoining the gas box, wherein the exit orifice has a controllable radius and length to allow plasma to flow out from the gas box to populate the SOL with the plasma. The system may also include fusion products, which decrease in speed in the plasma in the SOL, allowing energy to be extracted and converted into thrust or electrical power and further allowing ash to be extracted to reduce neutron emissions and maintain high, steady-state fusion power.
Cohen, Samuel A.↗
Thermal Gradient and Neutron Irradiation Experiment Design for Fusion Reactor Materials in the Advanced Test Reactor
This work outlines a hypothetical coupled thermal gradient and neutron irradiation experiment in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. Although the ATR is a thermal spectrum test reactor and doesn’t inherently produce a flux spectrum dominated by the high-energy neutrons typical in a fusion reactor, it’s multitude of experiment positions and dynamic flux environment make it a suitable platform for investigating fusion related issues.
36 MATERIALS SCIENCE↗
Multi-scale modelling to relate beryllium surface temperature, deuterium concentration and erosion in fusion reactor environment
Beryllium (Be) has been chosen as the plasma-facing material for the main wall of ITER, the next generation fusion reactor. Identifying the key parameters that determine Be erosion under reactor relevant conditions is vital to predict the ITER plasma-facing component lifetime and viability. To date, a certain prediction of Be erosion, focusing on the effect of two such parameters, surface temperature and D surface content, has not been achieved. In this paper, we develop the first multi-scale KMC-MD modeling approach for Be to provide a more accurate database for its erosion, as well as investigating parameters that affect erosion. First, we calculate the complex relationship between surface temperature and D concentration precisely by simulating the time evolution of the system using an object kinetic Monte Carlo (OKMC) technique. These simulations provide a D surface concentration profile for any surface temperature and incoming D energy. We then describe how this profile can be implemented as a starting configuration in molecular dynamics (MD) simulations. We finally use MD simulations to investigate the effect of temperature (300–800 K) and impact energy (10–200 eV) on the erosion of Be due to D plasma irradiations. The results reveal a strong dependency of the D surface content on temperature. Increasing the surface temperature leads to a lower D concentration at the surface, because of the tendency of D atoms to avoid being accommodated in a vacancy, and de-trapping from impurity sites diffuse fast toward bulk. At the next step, total and molecular Be erosion yields due to D irradiations are analyzed using MD simulations. The results show a strong dependency of erosion yields on surface temperature and incoming ion energy. The total Be erosion yield increases with temperature for impact energies up to 100 eV. However, increasing temperature and impact energy results in a lower fraction of Be atoms being sputtered as BeD molecules due to the lower D surface concentrations at higher temperatures. Finally, these findings correlate well with different experiments performed at JET and PISCES-B devices.
36 MATERIALS SCIENCE↗
Impact of fusion reactor neutronics modeling for transmutation and thermal feedback *
Abstract Fusion neutronics calculations provide important metrics pertinent to fusion device operations, such as tritium breeding ratios (TBRs) and data on heat deposition, material activation, and damage. Because of the high computational burden required to generate a high-fidelity Monte Carlo simulation of a 3D fusion device, various assumptions are made to reduce computational time by simplifying the reactor model or the calculation iteration. This paper explores the impact of fusion neutronics metrics such as the TBR and decay heat of structural materials based on assumptions of material composition in the fusion reactor and temperature modeling of materials. Results show that for compact tokamaks with high power and long operational cycles, the transmutation of structural materials is significant enough to cause a substantial change in the flux spectrum and decrease the TBR by 1.68% after 2 years of full power operation. Additionally, assuming a constant temperature and material density can impact the TBR calculations up to 3%.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
FY25 Theory and Simulation Performance Target: Development of an integrated modeling framework for fusion reactor design and assessment (Final Report)
This report documents the FY25 Theory and Simulation Performance Target (TSPT) of developing an integrated modeling framework for fusion reactor design and assessment (FREDA). Over Q1-Q4, new capabilities were developed across both plasma and engineering domains and demonstrated on an example representation of a Compact Advanced Tokamak with a Dual Cooled Lead Lithium blanket. This represents a first-of-a-kind demonstration of coupled core-to-wall-to-engineering for a reactor. Self-consistent CESOL workflows were applied to provide core, pedestal, and SOL prediction; new modules were developed for energetic particle stability (FAR3D) and transport (TGLF-EP) analysis; and boundary plasma modeling (SOLPS-ITER, BOUT++/Hermes-3) was expanded to evaluate wall and divertor heat fluxes and interface with engineering thermal analysis. A parameterized CAD tool, TRACER, was expanded to generate medium-fidelity divertor, blanket, and coil geometries; OpenFOAM and Diablo workflows were applied for first-wall and divertor thermal analyses with helium cooling; and reduced-order models were created for high-mass-flux divertor cooling. Magnet multiphysics capabilities were verified between Elmer, Diablo, and a new MFEM-based solver, and workflows enable stress, thermal, and neutron-fluence analysis of TF coils with neutronics-driven heating. Nuclear and blanket analysis workflows were demonstrated, including tritium breeding, transport, and CFD-informed thermo-mechanical assessment. Preliminary multi-fidelity uncertainty quantification workflows were applied to boundary modeling codes and shown to achieve variance reductions with fewer high-fidelity boundary simulations. Key findings highlight the challenges of resolving the ITEP gap to find suitable balance between wall and divertor loads, neutron heating, and practical limits of PFC cooling. Next step priorities are to develop automated workflows to check boundary code convergence and detachment, implement tighter physics-engineering CAD provenance tracking, and inclusion of plasma-material interface models for SLAG and tungsten cracking behavior. Collectively, these developments establish sophisticated capabilities for predictive, multi-fidelity, whole-device modeling that integrates plasma physics, materials, magnets, and nuclear engineering to guide pathways to viable Fusion Pilot Plant design points.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Deuterium retention in advanced steels for fusion reactor structural application
Reduced activation ferritic-martensitic (RAFM) steels, castable nanostructured alloys (CNA), and oxide-dispersion-strengthened (ODS) steels have been developed for the fusion reactor structural application with improved mechanical properties and radiation resistance. However, the hydrogen isotope retention in these steels is not well understood. In this study, 10 keV D 2 implantation was performed on six steels, together with pure iron and T91 as references, at room temperature. Subsequent thermal desorption spectroscopy (TDS) measurements were carried out to investigate the deuterium retention. Possible deuterium trapping sites in the studied materials were discussed in coordination with available microstructural information. The results of this study indicated that ODS-steels have the largest deuterium retention, which is almost one order of magnitude higher than that of pure iron, ~6 times of RAFM, and 2–4 times of CNA steels. The difference in nanoparticles and grain/lath boundaries are primarily responsible for the different deuterium retention, while other trapping sites (e.g., vacancy clusters, dislocations, etc.) are present. Deuterium retention increases with increasing sink strength of the studied materials, although they are not linear correlated.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Accelerated kinetic model for global macro stability studies of high-beta fusion reactors
The field reversed configuration (FRC), such as studied in the C-2W experiment at TAE Technologies, is an attractive candidate for realizing a nuclear fusion reactor. In an FRC, kinetic ion effects play the majority role in macroscopic stability, which allows global stability studies to make use of fluid-kinetic hybrid (also referred to as Ohm's law) models wherein ions are treated kinetically while electrons are treated as a fluid. The development and validation of such a hybrid particle-in-cell algorithm in the Exascale Computing Project code WarpX are reported here. Implementation of this model in the WarpX framework benefits from the numerical efficiency of WarpX as well as its scalability on large HPC systems and portability to different architectures. Performance benchmarks of the new algorithm for large, 3-dimensional, full device simulations from the Perlmutter supercomputer are presented. Results of a series of FRC simulations are discussed in which the impact of two-fluid effects on the tilt-mode growth rate was studied. It was observed that, in agreement with previous Hall-MHD studies, two-fluid effects have a stabilizing impact on the tilt mode.
Physics↗